IR 05000344/1987043
| ML20235A181 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 12/23/1987 |
| From: | Clark C, Mendonca M, Pereira D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20235A176 | List: |
| References | |
| 50-344-87-43, IEIN-87-035, IEIN-87-35, NUDOCS 8801120141 | |
| Download: ML20235A181 (8) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report No:
50-344/87-43 Docket No.
50-344 License No. NPF-1 Licensee:
Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name:
Trojan Nuclear Plant Inspection at:
Rainier, Oregon
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Inspection conducted:
December 7-11, 1987
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Inspector:
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. ark,.Reac or Inspector Dete Si ned
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Inspector:
zen de
/2 f d'7 D. B. Pereira, Reactor Inspector Date Signed Approved by:
%% %Me x
/z/r_s/s7 M. M. Mendonca, Chief, Date Signe!
Reactor Project Section 1 Summary:
Inspection During the Period of December 7-11 1987 (Report 50-344/87-43)
Areas Inspected:
This routine, unannounced inspection by the Project Inspector involved the areas of Temporary Instruction 25".5/84 Verification of Compliance with Order for Modification of License:
Prirary Coolant System Pressure Isolation (Event V) Valves, Followup items, anJ Onsite Followup of current events.
During this inspection, inspection modules 30703, 25534, 92701, 30702, 90713 and 92700 were used.
Results: No' violations or deviations were identified.
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8801120141 871223 PDR ADOCK 05000344 O
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DETAILS 1.
Persons Contacted a.
Licensee Personnel
"C. A. Olmstead, Plant Manager R. P. Schmitt, Manager, Operations and Maintenance
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I J. D. Reid, Manager, Plant-Services
- M. G. Cooksey, Supervisor, Control and Electrical Maintenance
- D. Nordstrom, Engineer, Nuclear Safety and Regulation Department
- J. K. Aldersebaes, Manager, ' Plant Modifications R. Reinart, Supervisor,. Instrument and Control b.
U. S. Nuclear Regulatory Commission
- R. Barr c.
Oregon Department of Energy
- H. Mooney, Oregon Resident Inspector
Attended the Exit Meeting on December 11, 1987.
2.
Temporary Instruction 2515/84 Verification of Compliance With Order For Modification of License : Primary Coolant System Pressure Isolation (Event V) Valves The purpose of tnis temporary instruction is to provide inspection requirements and quidance for followup of licensee's actions taken in response to the April 20, 1981 Order for Modification of License concerning Primary Coolant System Pressure isolation Valves. This temporary instruction (TI) is intended to verify satisfactory completion of lic-ensee actions in the implementation of periodic Event V valve testing in response to multiplant action (MPA) B-45.
As a matter of background, the Reactor Stafe; Study (RSS), WASH-1400, identified in a PWR an intersystem loss of coolant accident (LOCA) that is a significant contributor to risk of core melt accidents (Event V).
The design examined in the RSS contained in-series check valves isolating the high pressure primary coolant system from the low pressure injection system piping.
The scenario which leads to the Event V accident is initiated by the failure of these check valves to function as a pressure isolation barrier against reactor coolant system pressure.
This causes an overpressurization and rupture of the low pressure piping which results in a LOCA that bypasses containment.
To better define the Event V concern, all light water reactor licensees were requested by letter dated February 23, 1980, to provide the i
following in accordance with 10 CFR 50.54(f):
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a.
Describe the valve configurations and indicate if an Event V isolation valve configuration existed within the Class 1 l
boundary of the high pressure system piping, e.g., (1) two
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check valves in series, or (2) two check valves in series with an open motor-operated valve.
b.
If either of the above Event V configurations existed, indicate whether continuous surveillance or periodic tests were being performed on such valves to ensure integrity..Also indicate whether valves had been known, or found, to-lack integrity.
c.
If either of the above Event V configurations existed, indicate whether plant procedures should be revised or if plant medications should be made to increase reliability.
In addition to the above, licensees were asked to perform individual
check valve leak testing before plant startup after the next scheduled I
outage.
Based on licensee responses and the ongoing unsatisfactory operational experience at several plants, it was concluded that a valve configuration of concern existed at 36 plants.
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It was, therefore, apparent that when pressure isolation is provided by
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two in-series check valves and when failure of one valve in the pair can l
go undetected for a substantial length of time, verification of valve integrity is required.
Since these valves are safety-related, they must be tested periodically to ensure low probability of gross failure.
As a
result, it was determined that periodic examination of check valves was i
required to be undertaken by the licensee to verify that each valve is seated properly and functioning as a pressure isolation device.
Such testing was intended to reduce the overall risk of an inter-system LOCA.
On April 20, 1981, an order requiring the above described testing was sent to 32 PWR plants and 2 BWR plants.
This Order included a Safety Evaluation Report (SER) and Technical Specification insert pages to require leak rate testing of Event V pressure isolation vaives.
The requirements for TI 2515/84 were to review the documentation q
associated with the implementation of the Event V Order from'1980 to
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present for at least one pair of check valves.
Select two representative pairs of valves if two or more systems contain Event V valves.
The following inspection results present the documentation review as required
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by TI 2515/84.
The inspector reviewed the licensee's Technical Specifications (TS) to ensure that the modification was entered as required by the Event V Order.
Paragraph 4.4.6.2.2 was added under the Reactor Coolant System (RCS) Surveillance Requirements specifying each Reactor Coolant System Pressure Isolation Valve as specified in Table 3.4.6-1 to be demonstrated
operable by individually verifying leakage to be within its limit.
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addition, Table 3.4.6-1 was added specifying the valves to be leak tested and the maximum allowable leakage fcr each valve.
Both requirements were entered as required by the Event V Order dated April 20, 1981.
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The inspector verified that the test procedure reflected the requirements of the TS.
The licensee performs Periodic Operating Test (POT) 2-4, ECCS Pressure Boundary and Accumulator Valve Leakage Inservice Test which provides instructions for leak testing of the RCS pressure boundary check valves of the RHR, SIS and accumulator injection lines.
The test method used in POT 2-4 for measuring leakage is by pressure indicators initially and then using flow indicators if leakage is observed past the valve.
This is permissible in order to satisfy ALARA requirements, since leakage may be measured indirectly.
The inspector determined that this was an acceptable test method.
The inspector verified that the test procedure obtains the leakage rates for individual valves in some sections of POT 2-4, and combines the leakage rates for check valves that are in parallel.
Paragraphs 7.1, 7.4, and 7.8 of POT 2-4 perform individual leak tests for the RCS Cold
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legs, RCS Hot legs, and SI to RCS Hot legs check valves, respectively.
Paragraph 7.5 performs the leak test of the RHR to RCS Cold legs check valves (8818 A&B) together and then leak tests check valves (8818 C&D)
together.
Paragraph 7.6 leak tests the RHR to RCS Hot legs check valves (8736 A&B) together.
Paragraph 7.7 leak tests the SI to RCS Cold legs check valves (8819A-D) all together.
Since POT 2-4 Sections 7.5 through 7.7 test two or more valves in parallel, and if leakage exceeds 1.0 gpm, additional testing may be requested by the Plant Test Engineer to isolate the source of the leakage.
The inspector verified that the acceptance criteria stated in the test procedure are in accordance with the Technical Specifications.
If a leak test parameter falls outside the limits labeled (STS), the component shall be declared inoperable and the appropriate STS action statement applied.
The inspector verified that test data reflected the requirements as specified in the TS.
The inspector reviewed test records for past check valve leakage tests and observed the following:
a.
The test records contained major test data such as leak volume per unit time, upstream and downstream pressures, and leakage rate acceptance criteria.
b.
The recorded test frequency was in accordance with the TS.
c.
The as found leakage (i.e., prior to valve stroking, modification, adjustments, etc.) was recorded.
d.
Adequate corrective actions were taken for valves not meeting the acceptance criteria.
Where the leakage rate or trending criteria were exceeded, the licensee in the event of valve modification or reple, cement performed the required post-maintenance leakage rate testing.
The licensee in several examples initiated maintenance requests for check valve repair and performed the POT 2-4 retests satisfactorily, e.
The test data had no indication of anomalies which could indicate improper or inaccurate testing.
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The test data sheets appear to have an adequate review and approval process since the Shift Supervisor, the Plant Test Engineer, and the Engineering Supervisor have signature blocks at the end of the data
sheets.
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The test data records indicated that the test prerequisites as specified in the test procedure were satisfied, post-test activities (such as valve lineups and equipment readjustments), and I
documentation were completed correctly.
Based upon the review of the licensee's records, procedures and Technical Specification changes based upon the April 20, 1981 Order, it appears that the licensee has successfully implemented the periodic Event V check valve testing in response to multiplant action B-45.
The inspector verifies satisfactory completion and reports that TI 2515/84 is complete.
Temporary Instruction 2515/84 is closed.
No violations or deviations were identified in this area.
3.
Followup on Previous Inspection Findings a.
IE Information Notice No. 87-35 (Closed) Reactor Trip Breaker Westinghouse Model 05-416, Failed to Open on Manual Initiation From the Control Room Information Notice No. 87-35 was provided to alert licensees of a potentially significant safety problem associated with a reactor trip breaker (RTB), particularly a Westinghouse model D5-416.
A description of the circumstances of the safety problem was reported by the McGuire Nuclear Station Unit 2 on July 2, 1987 while performing control rod drop testing after a recent refueling outage.
The test was in progress with the plant in mode 3 (hot shutdown).
With all control rods inserted and the RTBs closed for testing the next bank of control rods, station personnel smelled smoke in the vicinity of the RTBs.
A manual trip of A and B train RTBs was initiated from the control room.
Only the A train RTB opened.
The B train RTB was eventually tripped manually at the breaker panel.
The smoke had come from the B train breaker shunt trip coil, which had burned and shorted while trying to open the breaker.
The coil is designed for intermittent duty and to carry current only until the breaker opens.
Failure of the breaker to open resulted in a prolonged and damaging current.
Operators in the control room stated that open indications for both the A and B train redundant RTBs were observed for all attempted breaker opening evolutions during the control rod drop testing process.
However, the event recorder indicated that the B train RTB failed to open on a previous manual trip attempt (approximately 4 minutes before) when operators were setting up for the control rod drop test on the last bank of rods.
An NRC Augmented Inspection Team (AIT) evaluated the licensee's investigation into the reactor trip breaker problem and discovered abnormal wear and a broken weld were found in this early vintage of Westinghouse DS-416 breaker.
The broken weld was on the main drive
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link between the center pole lever and the pole shaft.
Except for the shunt trip coil that had burned and shorted while trying to open the breaker, the breaker's electrical controls and auxiliary contracts were verified to be properly wired and operating as designed.
The cause for the anomalous breaker status indication is still under investigation.
Attempts to repeat the condition, where the breaker was mechanically binding in the closed position, were minimally successful.
Preliminary conclusions of the AIT are that the breaker's mechanical binding was caused by a combination of wear (greater than 2000 cycles of operation), manufacturing tolerances in this early vintage breaker, and the broken weld.
These factors may have combined to allow sufficient lateral movement of the main linkage to cause it to j
jam at or near full breaker closure and thus prevent the breaker
from opening.
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Trojan Nuclear Plant issued Operational Assessment Review (0AR)-87-55 to verify the nonapplicability of the specific 05-416 RTB problems to Trojan's DB-50 RTB's, and also to verify that the current Maintenance Procedure (MP) 1-5, entitled " Reactor Trip and Bypass Breakers" was adequate to detect similar mechanical problems in Trojan's 0B-50 RTB's.
Trojan's evaluation of 0AR 87-55 concluded the following:
1.
There are no 05-416 type circuit breakers in use at l
Trojan.
2.
The present MP 1-5 is adequate to ensure that a similar type event should not occur at Trojan.
The reasons for this rationale follows.
At McGuire, a broken weld on their 05-416 RTB permitted lateral movement of the main drive link which moved the roller close to its tolerance limits.
In the jammed position, the roller had slipped off the outer laminate of the cam.
The force exerted by the breaker closing action induced a twisting motion which
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caused the roller to wedge between the close cam J
lamination and the side frame.
Although it was established that the stacking of part tolerances played a part in the jamming of the breaker, it was also concluded
that the breaker would not jam unless a broken weld was present to permit the twisting action that allowed the roller to wedge.
3.
Although the general overall condition of Trojan's 08-50
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RTBs are checked in the present MP 1-5, which includes retaining rings, nuts, bolts, washers, freedom of movement, and functional testing, the documented inspection of welds is not performed.
Since, MP 1-5 will be revised to reflect the improvements and recommendations of the new Westinghouse DB-50 Maintenance Manual, as part of this revision, all visible welds will be inspected in
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the 08-50 RTB. MP 1-5 procedure change was in progress at the time of this inspection, j
Based on the Licensee's evaluation per their 0AR 87-55, and their proposed corrections to MP 1-5, the inspector considers Information Notice No.87-35 closed.
4.
Onsite Followup of Current Events
On Sunday, December 6, 1987, with the plant operating at 100% power, the turbine and reactor were manually tripped at 10:10 a.m. following a ramp load rejection from 1100 MWe to 20 MWe. The licensee's investigation indicated that the microswitch located in the control room on turbine d
control panel C05 for.the generator load decrease pushbutton had failed closed. The plant had just recovered from a 60 MWe grid initiated load
rejection at 9:50 a.m.
In the recovery process, the generator load l
increase and decrease pushbuttons had operated normally. The safety i
systems functioned as expected with the exception of the 'A' train of the
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auxiliary feedwater system. The trip and throttle valve, M0-3071,' for the steam driven turbine, was found to be in the closed positon and did not open on the automatic start signal. Troubleshooting revealed an open lead in the automatic' start circuitry.
On Monday, December 7, 1987, the licensee initiated a startup after completing repairs to the above problems, and experienced a blown fuse on the Rod Control System which delayed the startup and subsequent achieving power to go on the grid. Later in the day, when the licensee attempted to load the turbine with the generator load increase pushbutton, the generator did not load. Subsequent troubleshooting by the Instrumentation and Control (I&C) technicians and supervisors revealed
incorrect installation of a jumper across an Automatic Dispatch System (ADS) pushbutton which effectively bypassed the generator load increase pushbutton. This jumper was installed sometime during the troubleshooting phase on December 6, 1987 by revising the original Temporary Modification which removed the defective generator load increase pushbutton and installed the inactive ADS pushbutton. The I&C personnel performing the temporary modification, upon removal of the ADS pushbutton,-made a conscious decision to add a jumper between two leads from which the ADS pushbutton was just disconnected.
The Temporary Modification was incorrectly modified to add the jumper, even though reviewed by the I&C supervisor. This was a problem in lack of procedure compliance with the licensee's temporary modification procedure.
During backshift coverage, the Senior Resident Inspector observed the I&C personnel removing the' jumper which was the first hint of a problem in controlling modifications.
In addition, a complete and comprehensive test of the new installed generator load increase pushbutton would have prevented the delay and subsequent shutdown.
This inspector became involved in the sequence of events which eventually required the licensee to place an administrative hold on the startup and subsequent sorting of problems and errors generated. These topics will
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Review of Licensee Inservice Inspection Report An inspector reviewed the Trojan Unit 1 "1987 Inservice Inspection Report", which covered the 1987 refueling outage. Since the licensee extended their first ten year inspection interval by one year, this report constitutes the final inspection report for the first'10-year interval. The information provided in this report meets the reporting requirements of ASME Section XI, article IWA-6000 " Records and' Reports".
l No violation or deviations were identified.
6.
Exit Interview l
l The inspector met with the licensee representatives denoted in paragraph 1 on December 11, 1987, and summarized the scope and findings of the inspection activities.
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